Progress Towards High Performance, Steady-state Spherical Torus

Abstract

-Research on the Spherical Torus (or Spherical Tokamak) is being pursuedto explore the scientific benefits of modifying the field line structure from that in more moderate aspect ratio devices, such as the conventional tokamak. The ST experiments are being conducted in various US research facilities including the MAclass National Spherical Torus Experiment (NSTX) at Princeton, and three medium size ST research facilities: Pegasus at University of Wisconsin, HIT-II at University of Washington, and CDX-U at Princeton. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a Demo device, are being discussed. For these, it is essential to develop high performance, steady-state operational scenarios. The relevant scientific issues are energy confinement, MHD stability at high beta (b), non-inductive sustainment, ohmic-solenoid-free start-up, and power and particle handling. In the confinement area, the NSTX experiments have shown that the confinement can be up to 50% better than the ITER-98-pby2 H-mode scaling, consistent with the requirements for an ST-based CTF and Demo. In NSTX, CTF-relevant average toroidal beta values b T of up to 35% with the near unity central b T have been obtained. NSTX will be exploring advanced regimes where b T up to 40% can be sustained through active stabilization of resistive wall modes. To date, the most successful technique for non-inductive sustainment in NSTX is the high beta-poloidal -3 -regime, where discharges with a high non-inductive fraction (~60% bootstrap current + NBI current drive) were sustained over the resistive skin time. Research on radiofrequency based heating and current drive utilizing HHFW and EBW is also pursued on NSTX, Pegasus, and CDX-U. For non-inductive start-up, the Coaxial Helicity Injection (CHI), developed in HIT/HIT-II, has been adopted on NSTX to test the method up to Ip ~ 500 kA. In parallel, startup using RF current drive and only external poloidal field coils are being developed on NSTX. The area of power and particle handling is expected to be challenging because of the higher power density is expected in the ST relative to that in conventional aspect-ratio tokamaks. Due to its promise for power and particle handling, liquid lithium is being studied in CDX-U as a potential plasma-facing surface for a fusion reactor.